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  • Author or Editor: J. Ahn x
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Abstract  

The distillation behaviour of cadmium at a reduced pressure was investigated to develop an actinide recovery process from a liquid cadmium cathode in a laboratory scale cadmium distiller. The apparent evaporation rate of cadmium increased with an increasing temperature whereas the rate decreased with an increasing vacuum pressure. The evaporation rate of cadmium varied within 9.7–40 g/cm2/h in the temperature range of 500–650 °C and pressure range of 0.5–10 Torr (0.0667–1.33 kPa). The theoretical values calculated by the Hertz–Langmuir relation were much higher than experimentally measured values. The deviation was compensated by an evaporation coefficient (α) obtained empirically. About 0.02–0.20 wt% of residue was left in the crucible after distillation and found to be CdO. It could be concluded that the temperature range of 500–650 °C is favourable for the cadmium distillation process if residual eutectic salt does not exist in the cadmium alloy surface.

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Abstract  

It is important to increase a throughput of the salt removal process from uranium deposits which is generated on the solid cathode of electro-refiner in pyroprocess. In this study, it was proposed to increase the throughput of the salt removal process by the separation of the liquid salt prior to the distillation of the LiCl–KCl eutectic salt from the uranium deposits. The feasibility of liquid salt separation was examined by salt separation experiments on a stainless steel sieve. It was found that the amount of salt to be distilled could be reduced by the liquid salt separation prior to the salt distillation. The residual salt remained in the deposits after the liquid salt separation was successfully removed further by the vacuum distillation. It was concluded that the combination of a liquid salt separation and a vacuum distillation is an effective route for the achievement of a high throughput performance in the salt separation process.

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Abstract  

In order to be more confident of the performance assessment of high-level radioactive waste disposal, radionuclide transport must be investigated in more detail in the disturbed host rock region adjacent to the engineered barriers where disturbance has been introduced during the construction and waste-emplacement period. Geochemical, hydrological, and rock-mechanical properties should be quite different from those of undisturbed host rock. We have to elucidate the effect of bentonite intrusion into intersecting fractures from the standpoint of radionuclide confinement. In the present work, sorption distribution ratios (Kd's) of Np and Am are measured experimentally for various values or redox potential (Eh) in a simulated rock fracture filled with bentonite. The Kd of Am is approximately 6.5×103 ml/g and found to be insensitive to the redox potential. Under anaerobic conditions, the Kd of Np is approximately 6×104 ml/g. Under aerobic conditions, Kd is as small as 30 to 100 ml/g. This is the first report to measure the sorption behavior of Np and Am in a simulated rock fracture filled with bentonite (namely, in a disturbed zone) under pH, Eh and ionic strength control. We aan make use of these Kd data for numerically evaluating the mass transfer from bentonite filled fractures into the water-flowing fracture network1.

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Abstract  

A delayed neutron counting system has been implemented at the HANARO research reactor in 2007. Thermal neutron flux measured at the NAA #2 irradiation hole coupled to the delayed counting system, was higher than 3 × 1013 n cm−2 s−1. The delayed neutron counting system is composed of 18 3He detectors which are divided into three groups with six detectors and the collected signals of each group are processed to a digital signal. The count numbers were measured with the uranium mass by using NIST SRMs under fixed analytical condition and their correlation could be determined. Finally, delayed neutron activation analysis has been carried out for the determination of uranium mass fraction in the collected environmental samples.

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This study was aimed at evaluating the efficacy of different mineral separation procedures to validate the EN1788 (2001) European Union standard protocol for better identification of irradiated fish and shellfish. The silicate minerals were isolated with physical density separation method from two types of non-irradiated freeze-dried fish and shellfish that included Pacific saury (Cololabis saira), mackerel (Scomber japonicus), shrimp (Penaeidae metapenaeus), and mussel (Mytilus coruscus). Radiation-specific thermoluminescence (TL) peaks (glow curve 1) were observed between 150–250 °C. The peaks are typical for the irradiated food; despite the samples being not irradiated. Apparently it showed that the isolated minerals were contaminated with organic materials such as bone, etc. Acid-hydrolysis digestion was employed to remove the possible contaminants. The minerals obtained through alternative pre-treatment showed no TL curves in radiation specific temperature range. Moreover, acid hydrolysis extraction resulted in producing higher mineral yields and lower background luminescence. Results were also confirmed by calculating TL ratios (glow curve 1/glow curve 2) to confirm the irradiation history of samples. Furthermore, different time and temperature treatments on TL intensity of irradiated standard quartz (SiO2) minerals showed that the acid-hydrolysis can be adjusted to 50 °C and 3 h for better luminescence determinations.

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Summary  

It is impossible to detect 14C and 3H by direct methods such as γ-spectroscopy because they are pure b-emitters and thus they are classified as hard to measure nuclides (HTM). In this paper the analysis results of 14C and 3H in the low level radioactive wastes (LLWs), including spent ion exchange resin, evaporated bottom and sludge are presented. The LLWs were generated by three nuclear power plants (NPPs), in Korea all with pressurized water type reactors (PWRs). A simultaneous separation procedure for 14C and 3H in LLWs was established by wet oxidation-acid stripping. A liquid scintillation analyzer was used for the measurement of 14C and 3H. It was found that the recovery of 14C and 3H was 82-99 and 78-103%, respectively, by wet oxidation-acid stripping with diluted standard solutions. At the lowest injection of 14C and 3H, i.e., at 1.44 Bq for 14C and 1.22 Bq for 3H, the minimum detectable activity (MDA) of 14C and 3H was calculated as 0.88 and 0.78 Bq/g, respectively, for the minimum allowable sample weight, using wet oxidation and 16 wt% H2SO4 acid. By the wet oxidation-16 wt% H2SO4 stripping method no interfering nuclides were detected in the trapping solution of 14CO2 and the distillate of 3H. The activity concentration range of 14C in the analyzed samples, i.e., spent ion exchange resin, evaporated bottom and sludge, was 0.17-110,000, 8.4-1380 and 0.1-10,006 Bq/g, respectively, and that of 3H in the same was from no detectable to 769, 134-14,383 and 0.7-4820 Bq/g, respectively.

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Journal of Radioanalytical and Nuclear Chemistry
Authors: K. Kim, J. Bae, B. Park, D. Ahn, S. Paek, S. Kwon, J. Shim, S. Kim, H. Lee, E. Kim, and I. Hwang

Abstract  

A pyrochemical processing has become one of the potential technologies for a future nuclear fuel cycle. An integrated multi-physics simulation and electrotransport model of a molten-salt electrolytic process are proposed and discussed with respect to the recovery of pure uranium when using thermochemical data. This study has been performed to provide information for diffusion boundary layers between the molten salt (KCl-LiCl) and electrode. The diffusion-controlled electrochemical model demonstrate a prediction of the electrotransport behaviors of LWR spent fuel as a function of the time up to the corresponding electrotransport satisfying a given applied current based on a galvanostatic electrolysis.

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